The present invention relates to nuclear boiling water (BWRs) and more particularly to containment utilizing lateral vents.
Existing large BWRs are of the forced-circulation type. In BWRs undergoing power generation operations, reactor coolant, initially in the form of sub-cooled liquid (e.g. water), is circulated by main coolant recirculation devices (e.g. jet pumps or mixed-flow motor-driven pumps) around a path a portion of which is comprised of the core lower plenum region (located at the bottommost section of the reactor), thence through the nuclear core and into a core upper plenum in communication with the core. Flow exiting the core upper plenum then passes through standpipes that lead to an assembly of steam separators. The reactor coolant exiting the nuclear core and passing into the core upper plenum is a two-phase mixture of steam and water, the proportion of which varies depending upon such factors as the power output from the fuel bundles, the amount of sub-cooling present in the coolant entering the fuel bundle, and the amount of flow through the bundles. This last factor depends on the power of the recirculation pumps and the hydrodynamic flow resistance presented by the fuel bundle geometry and wetted surfaces, and the amount of orificing representing restrictions to flow just prior to the coolant's entrance into the core fuel assembly.
Joining with the core effluent in the core upper plenum is the core "by-pass" flow, which is reactor coolant that has flowed from the core lower (entrance) plenum into the region external to the fuel assembly channels (but inside the core shroud), thence upwardly generally through the region occupied by cruciform-shaped control blades which stand in various degrees of insertion into the core, thence across the upper grid member (called the "top guide") which with its lattice-like configuration keeps the fuel assemblies in a regular array, and finally into the core upper plenum. This by-pass coolant stream at its discharge into the core upper plenum is compromised substantially of saturated liquid, with perhaps a small amount of steam. Within the core upper plenum, these two effluents--by-pass flow and fuel bundle exit flow--rapidly mix together and quickly lose identity from their origins.
Mechanical steam separation can be utilized to accomplish the separation of the steam from the steam/water mixture exiting the core. Some earlier BWR designs used free-surface steam separation where, just as in the household tea kettle, steam separates unaided from the free-surface, and saturated water remains in the bulk coolant, which in BWRs is recirculated back down the downcomer annulus. This type of steam separation is feasible so long as the steam-leaving velocity, i.e. the bulk average velocity of the steam taken across the available pathway flow area, is not large, i.e. is no greater than about 1.8 foot/second. If steam-leaving velocities exceed this value, there tends to be carried along with the steam an unacceptably high moisture content. The high moisture levels saturate the moisture-drying abilities of the steam dryer, thus resulting in an unacceptably high moisture content in the steam leaving the reactor and supplied to the turbine. When steam moisture contents are too high in the turbine steam flow, accelerated erosion can occur on first-stage turbine blades and the efficiency of the turbine is reduced.
It is possible to obtain free-surface separation capabilities if the reactor pressure vessel (RPV) cross-sectional area is made sufficiently large. However, cost economies dictate that minimum diameter RPVs be used, so that mechanical steam separation has been developed to handle the high power output steam production levels of modern BWRs. In these latter designs, the steam bulk average velocity moving through the wet steam plenum region immediately downstream of the mechanical steam separators is about 5 feet/second.
The fuel assemblies grouped over the central region of the core tend to have higher exit steam qualities than do bundles located at the peripheral region of the core. It is desirable, nonetheless, that the flow rates and steam/water mixture proportions entering the steam separator standpipes be relatively uniform. To facilitate gaining more nearly uniform steam/water mixture for entry into the standpipes, the standpipe entrances are separated from the fuel assemblies by a distance of, for example, about 5 feet. Turbulent mixing occurring between the plumes leaving adjacent fuel assemblies, each with a different void content, is one mechanism acting to produce a more nearly uniform mixture which enters into the steam separator standpipes. More important to achieving flow mixture uniformity, however, is the hydrodynamic flow resistance represented by the standpipes, each with their end-mounted steam separators. Complete flow mixture uniformity entering the standpipes is at best difficult to achieve and, even with a five-foot separation between fuel assembly exits and standpipe entrances, it is not a design basis used for reactor performance evaluations.
The steam separator assembly consists of a domed or flat-head base on top of which is welded an array of standpipes with a three-stage steam separator, for example, located at the top of each standpipe. One function of the standpipes is to provide a stand-off separation of the larger-diameter steam separators, which are generally arranged in a particularly tightly-compacted arrangement in which external diameters of adjacent separators are nearly touching with each other, so that separated liquid coolant discharged at the bottom of the separator has a more "open" flow path outwardly from the reactor longitudinal axis and out to the downcomer annulus region which lies at the inboard periphery to the RPV. A second purpose for the standpipes is a high-power-output natural-circulation reactor using mechanical steam separators is to provide juxtaposed regions which promote natural-circulation by means of a vertical region of two-phase (and, thus, low-density) coolant inside the standpipes which is juxtaposed against single-phase liquid coolant outside the standpipes in a so-called "downcomer region", in which region height provides a very significant part of the total natural circulation driving head for coolant flow circulation within the reactor.
The steam separator assembly rests on the top flange of the core shroud and forms the cover of the core discharge plenum ("core upper plenum") region. The seal between the separator assembly and core shroud flange is a metal-to-metal contact and does not require a gasket or other replacement sealing devices. The fixed axial flow type steam separators have no moving parts and are made of stainless steel, for example, to resist corrosion and erosion.
In each separator, the steam/water mixture rising through the standpipes (the "standpipe region") impinges upon vanes which give the mixture a spin, thus enabling a vortex wherein the centrifugal forces separate the water from the steam in each of three stages. Steam leaves the separator at the top of this assembly and passes into the wet steam plenum below the dryer. The separated water exits from the lower end of each stage of the separator and enters the pool (the "downcomer region") that surrounds the standpipes to join the downcomer flow. The steam exiting from all separators either may be in the same horizontal plane, or the separators may be arranged in a slightly crowned fashion at the center to compensate for the crowned water gradient of the pool surrounding the standpipes.
The steam separator assembly may be bolted to the core shroud flange by long hold-down bolts, or the separator together with the dryer assembly may be held down onto the core shroud flange by contact from the reactor head when the latter is assembled to the reactor vessel. The nominal volumetric envelope of the steam separator assembly is defined by the horizontal plane of its lower flange that contacts the core shroud flange, its cylindrical sides that provide part of the five-foot stand-off from the fuel assembly exits, the circumscribed diameter of the outermost row of standpipes, the circumscribed diameter of the outermost row of steam separators, and the generally horizontal plane of the exits to the steam separators.
The core upper plenum region in a BWR currently under design known as the "simplified boiling water reactor" (SBWR) is substantially devoid of other mechanical devices, pipes, or structures; whereas the core upper plenum of a BWR/6 and "advanced boiling water reactor" (ABWR) reactor design generally contains spargers and nozzles for core sprays, and distribution headers for core flooders, respectively. In both reactor types, these spargers/headers are located at the outer periphery of the core upper plenum, mounted below the core shroud flange so that the sparger/header is clear of the refueling removal path of peripheral fuel assemblies and, thus, do not become removed during core refueling operations.
With specific reference to a natural-circulation SBWR, it will be observed that there are no recirculation pumps to aid in coolant recirculation. Steam generation in the core produces a mixture of steam and water which, because of steam voids, is less dense than saturated or sub-cooled water. Thus, the boiling action in the core results in buoyancy forces which induce core coolant to rise upwardly, to be continuously replaced by non-voided coolant arriving from beneath the core in the core lower plenum region. As the coolant leaves the core, it rises through the core upper plenum region, then through the standpipes region, and finally into the steam separators. This voided mixture inside these standpipes continues to be less dense than non-voided coolant external to the standpipes, resulting in the development of additional buoyancy force to further drive the coolant circulation. That this process is quite effective in promoting coolant recirculation can be noted from reported tests made in forced-circulation power reactors where the coolant circulation pumps are shut off. Even with their relatively short steam separator standpipes, reactor power levels of 25% and coolant flow rates of 35% of rated flow, are readily and safely maintainable.
The SBWR reactor is but modestly different from the forced-circulation BWR, with the most prominent differences being that the standpipes region is to be considerably longer in the SBWR (to develop a higher differential head), the core overall height may be somewhat shorter (for example, being 8 or 9 feet active fuel length versus 12.5 feet active fuel length in recent forced-circulation reactors), and the core power density will be somewhat lower. The severity of orificing--a means to promote hydrodynamic stability--at the entrance to the BWR fuel bundles may be lessened. The fuel bundle may have a larger diameter fuel rod in, for example, a 6.times.6 rod array, whereas the rod array for a forced-circulation reactor often is an 8.times.8 rod array. The design flow rates per fuel bundle, and the flow rates per steam separator, will be somewhat reduced in the SBWR design. Fuel exit steam quality will be approximately the same between the two designs. In the SBWR reactor design, no spargers or discharge headers are installed in the core upper plenum, while in the ABWR reactor, spargers or discharge headers are installed in the upper core plenum.
In some versions of SBWR reactors under study, the standpipes are very long while the core upper plenum is short. In other versions, the converse is true. The present invention is applicable equally in either version.
With respect to safety aspects of BWRs, the most serious credible reactor accident is in general conceived as a rupture of the reactor pressure vessel (RPV) or of a major coolant line connected to the vessel. Such an occurrence is known as a loss of coolant accident (LOCA). To prevent the release of toxic products resulting from such an accident, the RPV is placed within a series of containment structures. BWRs have a primary and a secondary containment structure. The primary containment vessel consists of a drywell and a wetwell. In a majority of BWRs operating in the 1970's, the drywell is a steel pressure vessel shaped like an electrical light bulb. It is designed for a pressure of 350 kPa(g) and is tested above 420 kPa(g). The steel vessel is enclosed in a thick, reinforced concrete structure which provides the mechanical strength and also serves as a radiation seal. The drywell contains the reactor and the coolant recirculation pumps. The secondary containment vessel or shield building commonly is a rectangular structure of reinforced concrete about 1.0 m thick.
In more recent BWRs, the drywell is a concrete cylinder with a domed top. The wetwell is an annular chamber in which the water is retained by an interior rear wall and by the steel cylinder that is the primary containment structure. Connection between the drywell and the wetwell is provided by a number of horizontal cylindrical vents in the lower part of the drywell wall. A reinforced concrete shield building constitutes the secondary containment.
During a LOCA, the steam released by the flashing of the coolant water would be forced into the water of the wetwell and be condensed, thereby lowering the temperature and pressure of the drywell atmosphere. Hence, the wetwell commonly is referred to as the pressure suppression pool.
The development of vertical layer lateral vents for the pressure suppression pool is disclosed in U.S. Pat. No. 3,115,450. Such lateral vent concept allows a gradual increase in the air clearing load to the pressure suppression pool. In the SBWR, and possibly larger BWRs with passive features, there will be an advantage in using the heat sink offered by the several millions of kilograms of water comprising the suppression pool for the long term cooling of the containment. In the SWBR, long term heat removal is assured by the isolation condensers, but they require some bleeding to the pressure suppression pool to remove non-condensable gases that can otherwise accumulate in the isolation condensers, reducing their heat transfer capabilities. The outlet of this bleedline must be less submerged in the pressure suppression pool than the elevation of the uppermost horizontal vent on the drywell side of the drywell-wetwell boundary. This feature allows the pressure difference between the drywell and wetwell to drive any steam plus non-condensible mixture through the isolation condensers and to drive any residual steam vapor plus non-condensibles downstream into the wetwell. The pressure in the drywell of the BWR containment may become sub-atmospheric when cold water is injected into the RPV and the RPV overflows or the water spills out through the break. The containment liner, usually made of thin steel plates welded together and anchored to the containment wall, will not withstand negative pressures and will fail. Conventionally, a vacuum breaker is installed between the wetwell and the drywell which consists of a check valve which opens at a predetermined pressure differential, e.g. 4 kPa. However, there is a potential danger that this check valve will stay open. The envisaged design of the isolation condenser of the SBWR is dependent upon a higher pressure in the drywell than in the wetwell in order for non-condensibles to be transported by the bleedline to the wetwell. There is no pressure differential between the drywell and the wetwell with a vacuum breaker valve stuck in an open position. Also, there is no head between the isolation condenser and the wetwell, and correspondingly, no transport of non-condensibles to the wetwell. Gas blanketing of the isolation condenser cannot be excluded, as the non-condensibles accumulate in the isolation condenser. This will result in insufficient heat removal with consequent possible failure of the containment.